Liquid wastes from nuclear fuel reprocessing are extremely hazardous and expensive to store or dispose of Cesium-137, Strontium-90, actinides, transuranium elements (TRU), rare earth elements, and technetium, which are present in these wastes are of particular concern. The transuranium elements are typically alpha-emitters with very long half-lives, and cesium and strontium are the major heat generators of the waste and produce gamma and beta radiation. Technetium, as the pertechnetate ion, is very mobile in the environment and has an extremely long half life. To increase safe handling of the majority of the waste, and to significantly reduce the storage and/or disposal cost of the waste, it is desirable to partition the waste into two fractions. One containing the majority of the highly radioactive components, and one containing the bulk of the non-radioactive portion of the waste.
Nuclear waste exists in numerous forms and locations world-wide. The largest inventory of highly radioactive materials is produced from the reprocessing of spent nuclear fuel. The fission process produces a number of undesirable, highly-radioactive elements which accumulate in the nuclear fuel. For the reuse or recycling of the unused fissionable material left in the fuel, normally either uranium-235 or plutonium-239, a separation process is employed to partition the fissionable material from the undesirable fission products. This is normally accomplished by leaching or dissolution of a portion or all of the spent nuclear fuel material, followed by chemical separation. Early chemical separation processes were based on precipitation, where, for example, BiPO.sub.4 was used to coprecipitate plutonium for weapons-grade plutonium production. More recently, and by far the most common, solvent extraction processes utilizing tri-n-butyl phosphate are used to chemically separate uranium and/or plutonium from solutions resulting from dissolution or leaching of spent nuclear fuel. The remaining acidic liquid waste, containing the highly radioactive fission products and trace transuranic elements, has been accumulated and stored in various forms around the world for the past 45 years.
Facilities in the United Kingdom, France, Japan, Russia, and China currently use the PUREX process (plutonium-uranium extraction) to recover and purify uranium and/or plutonium. The United States utilized this process (or variations of this process) for commercial-fuel reprocessing at the West Valley Plant, in upstate New York in the 1970's, and for reprocessing of weapons-grade plutonium at the Hanford Site in eastern Washington State, aluminum driver fuel at Savannah River Site in South Carolina, and naval fuel at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory (INEEL) in eastern Idaho. The reprocessing activities at Hanford were discontinued in the late 1980's and operations at Savannah River and the INEEL were discontinued in the early 1990's. There is currently no active nuclear fuel reprocessing facility in the United States.
There is however, a significant legacy of nuclear waste in storage from previous reprocessing activities. The majority of this waste was neutralized with caustic to facilitate storage in carbon steel vessels (Hanford and Savannah River). At the Idaho Chemical Processing Plant, this waste was calcined in a fluidized-bed calciner at 500.degree. C., producing a granular solid. This solid calcine is stored in stainless steel bins inside concrete vaults. Currently about 4100 m.sup.3 of highly radioactive calcine is stored at the INEEL, and about 1.2 million gallons of acidic liquid waste.
Currently, separate technologies are required for actinides and fission products, and often times, separate processes may be required for specific radionuclides such as cesium, strontium and technetium.
This invention describes novel extraction processes that will readily meet current safety standards and that will effectively separate the above-mentioned radioactive elements from typical nuclear reprocessing wastes. The liquid waste can be effectively decontaminated to meet low-level waste standards in two sequential solvent extraction processes.
U.S. Pat. No. 4,749,518 (Davis et al.) teaches a method for reprocessing nuclear waste by extracting cesium and strontium with crown compounds and cation exchangers.
U.S. Pat. No. 5,603,074 (Miller et al.) teaches a method of recovering cesium and strontium from an aqueous solution using a cobalt dicarbollide derivative.
U.S. Pat. No. 5,666,641 (Abney et al.) discloses a method of recovering cesium and strontium from an aqueous solution with polymeric materials and derivatives of cobalt dicarbollide.
U.S. Pat. Nos. 5,666,642 and 5,698, 169 (Hawthorne et al) teach the extraction of cesium and strontium ions from aqueous solutions, including fission product waste, using substituted metal dicarbollide ions.
To date, the known technologies have focused on recovering cesium and strontium from radioactive active wastes using various solvent systems, and have used the PUREX (plutonium uranium reduction and extraction) process for extracting actinides. However, past technologies have not provided for the selective sequential recovery of cesium, strontium, rare earth elements, technetium and actinides from aqueous acidic radioactive waste, using non-aromatic solvent extracting agents and non-hazardous stripping agents.